The Path of Greatest Certainty
A Generation III+ nuclear power plant > Readers accustomed to Anglo-Saxon units can use the following table to convert the main units from the International Metric System. 1 meter (m) = 3.2808 feet = 39.370 inches 1 square meter (m2) = 10.764 square feet 1 cubic meter (m3) = 219.97 imperial gallons = 264.17 US gallons 1 kilogram (kg) = 2.2046 pounds 1 tonne (t) = 0.984 long ton 1 bar = 14.504 psi
> Conversion of temperature (°C into °F) Temp. °C x 9/5 + 32 = Temp. °F
> All pressures are expressed in absolute bar. > FOREWORD
The need to secure long-term energy supplies, stabilize energy costs and combat global warming, argues in favor of a wide and diverse energy mix. Against this backdrop, nuclear power, which is proving increasingly competitive, safe, reliable and environmentally friendly, has a vital role to play.
As a world expert in energy, AREVA creates and offers solutions to generate, transmit and distribute electricity; its businesses are long-term, and cover every area of civil nuclear power generation to meet electricity needs. This encompasses the front end of the fuel cycle (uranium mining, conversion and enrichment, nuclear fuel fabrication), reactor design, construction, maintenance and services, the back end of the fuel cycle (treatment, recycling, transport and logistics, clean-up), and finally transmission and distribution of electricity from the generator to the high and medium voltage grids.
The EPR™ reactor is AREVA’s Generation III+ PWR (Pressurized Water Reactor). It has been designed to satisfy the needs of electrical utilities for a new generation of nuclear power plants, offering increased levels of safety and competitiveness, and meeting more efficiently tomorrow’s energy requirements. The EPR™ reactor is already under construction in Finland, France and China, and is currently undergoing licensing or pre-licensing in the US and UK.
Generation III+… we’re making it happen…
I 01 The Path of Greatest Certainty
Energy supply certainty ➡ The EPRTM design relies on a sound and The EPR™ reactor is a 1,600 + MWe PWR. Its evolutionary design proven technology. is based on experience from several thousand reactor-years of ➡ Continuous in-house design and operation of Light Water Reactors worldwide, primarily those manufacturing cooperation for a better incorporating the most recent technologies: the N4 (Chooz B1-B2 optimization. and Civaux 1-2) and KONVOI (Neckarwestheim-2, Isar-2 and ➡ Emsland) reactors currently in operation in France and Germany Design and licensing, construction respectively. The EPR™ design integrates the results of decades and commissioning, operability and TM of research and development programs, in particular those car- maintainability of EPR units benefit from ried out by the CEA (French Atomic Energy Commission) and the the long lasting and worldwide experience German Karlsruhe research center. Through its N4 and KONVOI and expertise of AREVA. filiation, the EPR™ reactor totally benefits from an uninterrupted evolutionary and innovative process; it has a proven technology Safety first and foremost based on 87 PWRs built throughout the world. The EPR™ technology offers a significantly enhanced level of Thanks to a number of technological advances, the EPR™ reactor safety: major safety systems consist of four separate subsystems, is at the forefront of nuclear power plants design. Significant contin- each capable of performing 100% of the safety function. Moreover uous improvement has been incorporated into its main features: functional diversity ensures that in case of total loss of a safety • the reactor core and its flexibility in terms of fuel management, system, in spite of its redundancy, the safety function can be performed by another system. In order to achieve its high level of • the reactor protection system, safety, the EPR™ reactor also features major innovations, especially • the instrumentation and control (I&C) system, the operator friendly in further preventing core meltdown and mitigating its potential human-machine interface and computerised control room of consequences. The EPR™ design also benefits from enhanced the plant, resistance to external hazards, including aircraft crash and • the large components such as the reactor pressure vessel earthquake. Together, the EPR™ operating and safety systems and its internal structures, the steam generators and the provide progressive responses commensurate with the potential primary coolant pumps. consequences of abnormal occurrences. AREVA NP’s Saint-Marcel and JSPM plants (in Chalon and Jeumont) have gathered over thirty years of experience in the man- Project and licensing certainty ufacturing of nuclear heavy components and are keeping it up to The French-German cooperation set up to develop the EPR™ date. technology brought together, from the start of the project: • power plant vendors, Framatome and Siemens KWU (whose nuclear activities have since been merged to form Framatome ANP, now AREVA NP),
> Building on experience Enhanced safety level and competitiveness
Evolutionary development keeps references
N4 Solid basis of experience KONVOI with outstanding performance Chooz B1&2, Civaux 1&2. Neckarwestheim-2, Isar-2 and Emsland.
02 I • EDF (Électricité de France) and the major German utilities Predictable business performance presently grouped in E.ON, EnBW and RWE Power, The forthcoming generation of nuclear power plants will have • the safety authorities from both countries to harmonise safety to demonstrate its competitiveness also in deregulated electric- regulations. ity markets. The EPR™ design takes into account the expectations of utilities Thanks to an early focus on economic performance during its as stated by the “European Utility Requirements” (EUR). It complies design process, the EPR™ technology offers significantly reduced with the specific requirements formulated by the French and German power generation costs, about 20% lower than those of large safety authorities for the next generation of nuclear reactors. combined-cycle gas plants. The Finnish electricity utility Teollisuuden Voima Oy (TVO) signed This high level of competitiveness is achieved through: a contract with the AREVA and Siemens consortium to build a ➡ turnkey EPR™ unit at the Olkiluoto site in Finland. The construction a unit power in the 1,600 + MWe range, permit was obtained in February 2005. providing an attractive cost of the kWe installed, On January 23, 2007, EDF ordered AREVA’s 100th nuclear reac- tor, which is being built in France, on the Flamanville site. The ➡ a 36-37% overall efficiency depending construction permit was awarded on April 10, 2007. on site conditions (presently the highest On November 26, 2007, AREVA and CGNPC signed a contract for value ever for water reactors), the supply of two EPR™ Nuclear Islands on the new site of Taishan ➡ a design for a 60-year service life, in China in the context of a long-term cooperation agreement. ➡ an enhanced and more flexible fuel And on April 23, 2008, E.ON chose the EPR™ reactor as its refer- utilisation, ence design for the new NPPs in the United Kingdom. ➡ an availability design target above 92%. AREVA’s supply chain is integrated, comprehensive and time-tested, which confers certainty to its projects: Significant advances • expertise is all within the company, at all levels, for sustainable development • the procurement schedule is controlled internally and therefore The EPR™ reactor due to its optimized core design and higher more flexible, overall efficiency offers many significant advantages in favor of • specific customer or regulator requirements are more efficiently sustainable development, typically: addressed. • 7-15% saving on uranium consumption per ➡ The EPR™ power plant is under construction produced MWh, in Finland, France and China, and • 10% reduction on long-lived actinides is currently undergoing licensing or generation per MWh, through improved fuel pre-licensing in the US and UK. management, ➡ The integration of design and manufacturing • 10% gain on the “electricity generation” strengthens AREVA’s supply chain and, versus “thermal release” ratio, as a result, project certainty. (compared to 1,000 MWe-class reactors).
➡Energy supply certainty with the ➡Procurement certainty for critical evolutionary design, operational flexibility components directly sourced from and shortened outages. AREVA’s existing integrated facilities. ➡Engineering certainty for customers ➡Project certainty with ongoing building through evolutionary design. experience and established supply chain. ➡Licensing certainty with construction ➡Business performance certainty with an license obtained in France and in Finland, efficiency up to 37%, flexible fuel management, licensing process launched in the United low operational maintenance costs. States, the United Kingdom and China.
I 03 > INTRODUCTION
In a nuclear power plant, the reactor is the component where the heat, necessary to produce the steam, is generated by the fission of atomic nuclei. The steam produced drives a turbine generator, which generates electricity. The nuclear steam supply system is the counterpart of the coal, gas or oil-fired boilers used in fossil-fuelled power plants.
In Pressurized Water Reactors (PWR) such as the EPR™ power plant, ordinary (light) water is utilized to remove the heat produced inside the reactor core by the nuclear fission phenomenon. This water also slows down (or moderates) neutrons (the constituents of atomic nuclei that are released in the nuclear fission process). Slowing down neutrons is necessary to sustain the nuclear reaction (neutrons must be Steam generator moderated to be able to break down the fissile atomic nuclei). Pressurizer The heat produced inside the reactor core is transferred to the turbine Control through the steam generators. rod drive Only heat is exchanged between the mechanism reactor cooling circuit (primary circuit) Primary and the steam circuit used to feed the pump turbine (secondary circuit). No exchange of cooling water takes place. The primary cooling water is pumped through the reactor core and the tubes inside the steam generators, in four parallel closed loops, by coolant pumps powered by electric motors. Reactor Each loop is equipped with a steam core generator and a coolant pump. The reactor operating pressure and temperature are such that the Vessel cooling water does not boil in the primary circuit but remains in the liquid state. A pressurizer, connected to one of the coolant loops, is used to control the pressure in the primary circuit.
04 I Feedwater entering the secondary side ➡The following chapters provide of the steam generators absorbs the a detailed explanation and description heat transferred from the primary side of a nuclear power station based and evaporates to produce saturated on an EPR™ reactor. steam. The steam is mechanically dried inside the steam generators then delivered to the turbine. After exiting the turbine, the steam is condensed and returned as feedwater to the steam generators.
A generator, driven by the turbine, Transformer generates electricity.
Generator
High voltage electrical lines Feedwater pump Condenser
Primary system Secondary system: Cooling – Steam Reheater water – Water
I 05 > TABLE OF CONTENTS
page 08 page 30 EPR™ NUCLEAR ISLAND SAFETY
>EPR™ REACTOR LAYOUT >NUCLEAR SAFETY Three protective barriers >PRIMARY SYSTEM Defense in depth >REACTOR CORE >EPR™ SAFETY SYSTEMS >SYSTEMS Design choices for reducing the probability of accidents Chemical and volume control liable to cause core melt Safety injection/ Design choices for limiting the residual heat removal consequences of severe accidents In-containment refuelling water storage tank (IRWST) Emergency feedwater Essential Service Water Ultimate Cooling Water system Other safety systems Component Cooling Water Other systems Power supply Fuel handling and storage
>INSTRUMENTATION & CONTROL SYSTEM EPR™ I&C overall architecture Role of the I&C systems
06 I page 38 page 40 EPR™ REACTOR CONSTRUCTION PLANT OPERATION, MAINTENANCE & SERVICES >EPR™ REACTOR CONSTRUCTION TIME SCHEDULE An availability design target above 92% Design features A high level of operational manoeuvrability Construction and installation methods An enhanced radiological protection Major component manufacturing Plant services Commissioning tests
page 44
ENVIRONMENTAL IMPACT
Design Construction Operations Decommissioning
I 07 > EPR™ REACTOR LAYOUT page 10
> PRIMARY SYSTEM page 14
> REACTOR CORE page 18
> SYSTEMS page 20
CHEMICAL AND VOLUME CONTROL page 20
SAFETY INJECTION/ RESIDUAL HEAT REMOVAL page 21
IN-CONTAINMENT REFUELLING WATER STORAGE TANK page 22
EMERGENCY FEEDWATER page 22
OTHER SAFETY SYSTEMS page 22
COMPONENT COOLING WATER page 23
ESSENTIAL SERVICE WATER page 23
ULTIMATE COOLING WATER SYSTEM page 23
OTHER SYSTEMS page 23
POWER SUPPLY page 24
FUEL HANDLING AND STORAGE page 25
> INSTRUMENTATION & CONTROL SYSTEM page 26
EPR™ I&C OVERALL ARCHITECTURE page 26
ROLE OF THE I&C SYSTEMS page 28
08 I EPR™ NUCLEAR ISLAND
Civaux nuclear power plant, France (N4, 1,500 MWe)
I 09 ■ EPR™ REACTOR NUCLEAR ISLAND
EPR™ REACTOR LAYOUT
7
4 3 3 1 3
2 3
5 4
6
1 Reactor Building are separated within the building. Areas of high activity are separated from areas of low activity by means of shielding facilities. The lower The Reactor Building located in the center of the Nuclear Island houses part of the building houses the fuel pool cooling system, the extra the main components of the Nuclear Steam Supply System (NSSS) borating system, and the chemical and volume control system. The as well as the In-Containment Refueling Water Storage Tank (IRWST). redundant trains of these systems are in two separate divisions of the Its main function is to prevent the release of radioactive materials into building that are physically separated by a wall. the environment under all circumstances, including possible accident conditions. It consists of a cylindrical pre-stressed concrete inner containment with a metallic liner, surrounded by an outer reinforced 3 The Safeguard Buildings concrete shell. The four Safeguard Buildings house key safety systems such as the The main steam and feedwater valves are housed in dedicated Safety Injection System and the Emergency Feedwater System, and reinforced concrete compartments adjacent to the Reactor Building. their support systems. These safety systems are divided into four trains each of which is housed in a separate division located in one The primary system arrangement is characterized by: of the four Safeguard Buildings. • a pressurizer located in a separate area, The combined Low Head Safety Injection System and Residual Heat • concrete walls between the loops and between the hot and cold Removal System is arranged in the inner radiologically controlled legs of each loop, areas, whereas the Component Cooling and Emergency Feedwater • a concrete wall (secondary shield wall) around the primary system Systems are installed in outer areas classified as radiologically to protect the containment from missiles that could be caused by non-controlled. The Main Control Room is located in one of the the failure of pressurized equipment, and to provide shielding Safeguard Buildings. against radiation produced by the primary system.
2 Fuel Building 4 Diesel Buildings The Fuel Building, located on the same basemat that supports the The two Diesel Buildings house the four emergency Diesel generators, Reactor Building and the Safeguard Buildings, houses an interim two Station Black Out Diesel (SBO) Generators and their support fuel storage pool for fresh and spent fuel and associated fuel handling systems. The SBO are used to supply electricity to the safety trains equipment. Operating compartments and passageways, equipment in the event of a complete loss of electrical power. The physical compartments, valve compartments and the connecting pipe ducts separation of these two buildings provides additional protection.
10 I Water outfall
Switchyard
Water intake
Nuclear Island Turbine Island Balance of Plant
Quay
(Typical layout - can be adjusted depending on site conditions).
5 Nuclear Auxiliary Building 6 Waste Building Part of the Nuclear Auxiliary Building (NAB) is designed as a The Waste Building is used to collect, store and treat liquid and solid radiologically non-controlled area in which parts of the Operational radioactive waste. Chilled Water System are located. Special laboratories for sampling systems are located at the lowest level. The maintenance area and some setdown areas used during the refueling phase are arranged 7 Turbine Building on the highest level. All air-exhausts from the radiologically controlled The Turbine Building houses all the main components of the steam- areas are routed, collected and controlled within the Nuclear Auxiliary condensate-feedwater cycle. It contains, in particular, the turbine, Building prior to release through the stack. the generator set, the condenser and their auxiliary systems.
I 11 ■ EPR™ NUCLEAR ISLAND
➡ The EPR™ reactor layout offers unique • To withstand the impact of a large resistance to external hazards, especially aircraft, the Reactor Building, Spent Fuel earthquake and aircraft crash. Building and two of the four Safeguard • To withstand major earthquakes, the Buildings are protected by an outer entire Nuclear Island stands on a single shell made of reinforced concrete. thick reinforced concrete basemat. The other two Safeguard Buildings The building height has been minimised are geographically separated. Similarly, and heavy components and water tanks the Diesel generators are located in two are located at the lowest possible level. geographically separate buildings.
12 I The outer shell (in blue in the image) protects the Reactor Building, the Spent Fuel Building and two of the four Safeguard Buildings including the Control Room.
➡ The EPR™ Nuclear Island design offers • Maintenance requirements were major advantages to operators, especially systematically taken into account for radiation protection and ease at the earliest stage of the design. of maintenance. For example, large setdown areas have • The layout is optimized and based on the been established to make maintenance strict separation of redundant systems. operations easier for operating personnel.
I 13 ■ EPR™ NUCLEAR ISLAND
PRIMARY SYSTEM
➡
PRIMARY SYSTEM CONFIGURATION The increased free volume in the reactor pressure vessel, between the inlet/outlet nozzles and the top of the core, provides a higher The EPR™ primary system is of a well-proven 4-loop design to water volume above the core and thus additional time before core which the French 1,300 MWe and 1,500 MWe N4 reactors as well uncovering in the event of a postulated loss of coolant accident. This as the German KONVOI reactors belong. gives the plant operator more time to counteract such event. In each of the four loops, primary coolant leaving the reactor pressure This increased volume is also beneficial in shutdown conditions in the vessel through an outlet nozzle passes into a steam generator, which hypothetical case of loss of the Residual Heat Removal System transfers heat to the secondary circuit. The coolant then passes function. through a reactor coolant pump before being returned to the reactor pressure vessel through an inlet nozzle. Inside the reactor pressure Larger water and steam phase volumes in the pressurizer smooth vessel, the coolant flows downward in the annular space between the the response of the plant to normal and abnormal operating core barrel and the vessel, then it makes an upward U turn and flows transients, allowing extended time to counteract accident situations through the core to extract the heat generated by the nuclear fuel. and an increased equipment lifetime. A pressurizer, part of the primary system, is connected to one of the The larger volume of the steam generator secondary side results in four loops. Its main role is to maintain the primary pressure within an increased secondary water inventory and steam volume, which a specified range. offers several advantages: • during normal operation, transients are smoother and the potential The main components of the EPR™ reactor, reactor pressure vessel, for unplanned reactor trips is therefore reduced; pressurizer and steam generators are of a larger volume than similar components in previous designs, providing additional operational and safety margins.
Integration of design and manufacturing Customers benefit greatly from the fact that heavy com- ponent design and manufacturing activities are brought together within the AREVA group. The possibility, unique in the nuclear market place, of having a very close connec- tion between the two is important for project success. This organisation implemented by AREVA NP since many years, is a great advantage for utilities. It provides the responsive- ness needed to achieve optimized design, manufacturing, schedule and cost to obtain the best solutions. Cattenom, France (4 x 1,300 MWe): inside a reactor building.
14 I CHARACTERISTICS DATA Reactor coolant system Core thermal power 4,590 MWth* Number of loops 4 Nominal flow (best estimate) 28,315 m3/h Reactor pressure vessel inlet temperature 295.2°C Reactor pressure vessel outlet temperature 330°C Primary side design pressure 176 bar Primary side operating pressure 155 bar Secondary side Secondary side design pressure 100 bar Saturation pressure at nominal conditions (SG outlet) 77.2 bar Main steam pressure at hot standby 90 bar * Depending on specific customer requirements.
• in hypothetical steam generator tube rupture scenarios, the combination of a larger steam volume and the use of a setpoint for the safety valves of the steam generators above the safety injection pressure, prevent liquid from being released into the environment, reducing the potential for radioactivity release outside the reactor containment; • in case of a total loss of the steam generator feedwater supply, the increased mass of water in the secondary system extends the dry- out time sufficiently to enable operators to recover feedwater supplies or to apply other countermeasures. In addition, the primary system design pressure has been increased in order to reduce the actuation frequency of the safety valves, which is also an enhancement in terms of safety.
➡ The increased volume of the primary system is beneficial in smoothing the effects of many types of transients. ➡ The primary system design pressure has been increased to reduce the frequency of safety valve actuation. ➡ The management of steam generator tube rupture scenarios prevents any liquid release outside the reactor containment. ➡ The large steam generator secondary side water inventory increases the time Computer-generated image available to take action in case of assumed of the EPR™ primary system total loss of secondary feedwater.
I 15 ■ EPR™ NUCLEAR ISLAND
The design of the EPR™ components includes many improvements with respect to previous designs. The most noteworthy are summarized below.
REACTOR PRESSURE VESSEL Neutron reflector • The absence of penetrations through the RPV bottom head The Neutron reflector is a stainless steel structure, increases its resistance in postulated core meltdown accidents surrounding the core, made of rings piled up one on top of and removes the need for the corresponding in-service inspection the other. It is an innovative feature that gives significant and potential repairs. benefits: • A reduced number of welds and improved weld geometry ➡ by reducing the flux of neutrons escaping from the core, compared to existing French and German 4-loop designs decrease the reflector ensures that a greater neutron fraction is the need for in-service inspections, facilitate non-destructive available to take part in the chain reaction process. The examination and reduce the duration of inspections. result is improved fuel utilisation, making it possible to • The residual cobalt content of the stainless steel cladding is decrease the fuel cycle cost; specified at a low value of less than 0.06% to contribute to the reduction in the radiation source term. ➡ by reducing the flux of neutrons escaping from the core, the reflector protects the reactor pressure vessel against aging and embrittlement induced by fast- neutron fluence, helping to ensure the 60-year design life of the EPR™ reactor pressure vessel.
Reactor pressure vessel monobloc upper shell for the Olkiluoto 3 (Finland) EPR™ reactor pressure vessel.
STANDSTILL SEAL (REACTOR COOLANT PUMP) The shaft seals are backed up with a standstill seal that closes once the pump is at rest and all seals of the leak-off lines are closed. The standstill seal creates a sealing surface with a metal-to-metal contact ensuring the shaft tightness in case of: • simultaneous loss of water supplies from the Chemical and Volume Control System and the Component Cooling Water System used to cool the shaft sealing system, • cascade failure of all the stages of the shaft sealing system. This feature ensures that even in case of total station blackout or failure of the main seals no loss of coolant would occur.
16 I AXIAL ECONOMIZER (STEAM GENERATOR) SECTION A Double wrapper To increase the heat transfer efficiency, the axial economizer directs Pressure shell 100% of the cold feedwater to the cold leg of the tube bundle, and about 90% of the hot recirculated water to the hot leg. This is done Bundle wrapper Divider plate by adding a wrapper to guide the feedwater to the cold leg of the 10% recirculated water tube bundle and a partition plate to separate the cold leg from the hot 90% recirculated water leg. This design improvement increases the steam pressure by about 100% feedwater 3 bar compared to a conventional steam generator. Pressure shell Double wrapper AA PRESSURIZER Divider plate Bundle wrapper The pressurizer has a larger volume to smooth the operating Hot leg Cold leg transients in order to: • ensure the equipment 60-year design life, • increase the time available to counteract an abnormal operating situation. Maintenance and repair (safety valves, heaters) are facilitated and radiological doses are reduced. A dedicated set of valves for depressurising the primary circuit is installed on the pressurizer, in addition to the usual relief and safety valves, to prevent the risk of high pressure core melt accident.
Computer-generated image of the EPR™ pressurizer head with its safety and relief valves.
I 17 ■ EPR™ NUCLEAR ISLAND
REACTOR CORE
➡
The reactor core contains the fuel material in which the fission Core instrumentation reaction takes place, releasing energy. The reactor internal structures support the fuel assemblies, channel the coolant and The ex-core instrumentation is used to monitor the process to guide the control rods which control the fission reaction. criticality and the core reactivity change, it can also measure the core power. The core is cooled and moderated by water at a pressure of 155 bar and a temperature in the range of 300°C. The coolant contains The reference instrumentation used to monitor the power distribution soluble boron as a neutron absorber. The boron concentration in the in the core is an “aeroball” system. Vanadium balls are periodically coolant is varied as required to control relatively slow reactivity inserted in the core. Their activation level is measured, to give values changes, including the effects of fuel burnup. Additional neutron of the local neutron flux which are used to construct a three- absorbers (gadolinium), in the form of burnable absorber-bearing dimensional power map of the core. fuel rods, are used to adjust the initial reactivity and power The fixed in-core instrumentation consists of neutron detectors and distribution. Instrumentation is located inside and outside the core to thermocouples, that are used to measure the neutron flux distribution monitor its nuclear and thermal-hydraulic performance and to provide in the core and the temperature distribution at the core outlet. input for control functions. The in-core instrumentation is introduced through the vessel head. The main features of the core and its operating conditions have been Therefore, the bottom of the reactor pressure vessel is free from any selected to obtain not only high thermal efficiency of the plant and penetration. low fuel cycle costs, but also extended flexibility for different irradiation cycle lengths and a high level of manoeuvrability.
Isar-2 unit, Germany (KONVOI, 1,300 MWe): fuel loading operation.
18 I In-core instrumentation
12 lance yokes, CHARACTERISTICS DATA each comprising: Reactor core – 3 T.C core 1 TC upper 89 control Thermal power 4,590 MWth* outlet plenum assemblies – One probe Operating pressure 155 bar containing Active fuel length 4,200 mm 6 in-core Number of fuel assemblies 241 detectors 4 water – 3 or 4 aeroball level Number of fuel rods 63,865 probes Average linear heat rate 166.7 W/cm * Depending on specific customer requirements.
Typical initial core loading T.C G G G G G G G G
G G
G G G G
Ex-core G G G G Aeroball
In-core G G
G G G G G G G G
G High enrichment Medium enrichment TC: Thermocouple with gadolinium Low enrichment High enrichment without gadolinium
➡The EPR™ core is characterized by ➡The EPR™ core also offers significant considerable margins for fuel management advantages in favor of sustainable optimization. development in comparison with current PWR designs: ➡Several types of fuel management (fuel cycle length, IN-OUT/OUT-IN) are available • 7-15% saving on uranium consumption to meet utilities’ requirements. per produced MWh, ➡The main features of the core and its • 10% reduction on long-lived actinides operating conditions give competitive generation per MWh (through improved fuel management cycle costs. fuel management).
I 19 ■ EPR™ NUCLEAR ISLAND
SYSTEMS
CHEMICAL AND VOLUME CONTROL • Ensuring a high flow rate capability for primary coolant chemical control, purification, treatment, degassing and storage. The Chemical and Volume Control System (CVCS) performs • Injection of cooled, purified water into the Reactor Coolant Pump several operational functions. (RCP) seal system to ensure cooling and leaktightness of the seals • Continuous control of the water inventory in the Reactor Coolant and collection of the seal leakage flow. System (RCS) during all normal plant operating conditions, using • Supply of borated water to the RCS up to the concentration the charging and letdown flow. required for cold shutdown conditions for any initial condition. • Adjustment of the RCS boron concentration by injecting • Reduction of RCS pressure to Residual Heat Removal System demineralized or borated water for control of power variations and (SIS/RHRS) operating conditions, by diverting charging flow to during plant start-up or shutdown, or to compensate for core the auxiliary pressurizer spray nozzle to condense steam in the burnup. pressurizer. • Ensuring the continuous monitoring of the boron concentration of • Filling and draining of the RCS during shutdown. all fluids injected into the RCS, and controlling the concentration • Provision of pressurizer auxiliary spray, if the normal system cannot and nature of dissolved gases in the RCS by injecting hydrogen perform its function; provision of make-up to the RCS in the event into the charging flow and degassing the letdown flow. of loss of inventory due to a small leak. • Adjustment of the RCS water chemical characteristics by injection • Ensuring the feed and bleed function. of chemical conditioning agents into the charging flow.
Chemical and Volume Control System
Regenerative Letdown heat exchanger Auxiliary spray Pressure reducing station HP coolers CCWS CCWS PRT
M Sampling LOOP 2 M system
LOOP 1 Reactor vessel Coolant Charging line purification
M
LOOP 4 LOOP 3 Coolant Seal injection degasification
Gas waste processing system Coolant storage Volume control tank Boric acid make-up IRWST IRWST Water make-up Gas treatment
20 I N2 H2 SAFETY INJECTION/ • transfers heat continuously from the RCS or the reactor refueling RESIDUAL HEAT REMOVAL pool to the CCWS during cold shutdowns and shutdown for refueling, as long as any fuel assemblies remain inside the The Safety Injection System (SIS/RHRS) comprises the Medium Head containment. Safety Injection System (MHSI), the accumulators, the Low Head Safety Injection System (LHSI) and the In-Containment Refuelling In the event of an assumed accident the SIS, operating in RHR mode Water Storage Tank (IRWST). The system performs a dual function in conjunction with the CCWS and the Essential Service Water being used both in normal operating conditions (in RHR mode) and in System (ESWS), maintains the RCS core outlet and hot leg the event of an accident. temperatures below 180°C following a reactor shutdown. The system is designed on the basis of a quadruple redundancy, The four redundant and independent SIS/RHRS trains are arranged i.e. it consists of four separate and independent trains, thus satisfying in separate divisions in the Safeguard Buildings. Each train is the N+2 concept (i.e. the capability of performing its function despite connected to one dedicated RCS loop and is designed to provide a single failure on one train and the unavailability of another train due the necessary injection capability required to mitigate accident to maintenance). Each train provides the capability for injection of water conditions. This configuration greatly simplifies the system design. into the RCS via an accumulator, a Medium Head Safety Injection The design also makes it possible to have extended periods available (MHSI) pump, and a Low Head Safety Injection (LHSI) pump for carrying out preventive maintenance or repairs. For example, discharging through a heat exchanger at the pump outlet. preventive maintenance can be carried out on an entire safety train During normal operating conditions, operating in RHR mode, the system: during power operation. • provides the capability for heat transfer from the RCS to the Component Cooling Water System (CCWS) when the RCS temperature is less than 120°C,
SI/RHR System
– Four-train SIS RHR – In-containment refuelling SI water storage tank – Combined RHRS/LHSI
Hot legs
LHSI RHR LHSI RHR pump pump
Accumulators Accumulators LHSI RHR LHSI RHR pump pump
Cold legs MHSI MHSI pump pump MHSI MHSI pump pump IRWST IRWST
Division 1 Division 2 Division 3 Division 4
I 21 ■ EPR™ NUCLEAR ISLAND
In safety injection mode, the main function of the SIS is to inject EMERGENCY FEEDWATER water into the reactor core following a postulated loss of coolant The Emergency Feedwater System (EFWS) is designed to ensure accident in order to compensate for the consequence of such that water is supplied to the steam generators when the systems events. It is also activated following a postulated steam generator that normally supply feedwater are unavailable. tube rupture or following the loss of the secondary-side heat removal function. Its main safety functions are to: • transfer heat from the RCS to the secondary side via the steam The MHSI system injects water into the RCS at a pressure (92 bar generators, in the period before connection of the RHRS, following shutoff head) which prevents the system from opening the secondary any plant incidents other than those involving a significant breach side safety valves (100 bar) following a steam generator tube of the reactor coolant pressure boundary; this is done in leakage. The accumulators and the LHSI system also inject water conjunction with the discharge of steam to the main condenser (if into the RCS cold legs when the primary pressure is sufficiently low available) or otherwise via the Main Steam Relief Valves or Safety (accumulator: 45 bar, LHSI: 21 bar shutoff head). Valves; • ensure that sufficient water is supplied to the steam generators following a loss of coolant accident or a steam generator tube IN-CONTAINMENT REFUELING WATER rupture accident; STORAGE TANK (IRWST) • rapidly cool the plant down to LHSI conditions following a small The IRWST is a tank containing a large volume of borated water, loss of coolant associated with total MHSI failure, in conjunction which is able to collect water discharged inside the containment in with steam release to the main condenser (if available) or otherwise postulated accident conditions. via the Main Steam Relief Valves or Safety Valves. Its main function is to supply water to the SIS, Containment Heat The system consists of four separate and independent trains, each Removal System (CHRS) and Chemical and Volume Control System providing injection capability through an emergency feed pump that (CVCS) pumps, and to flood the corium spreading area in the event takes suction from an EFWS tank. of a hypothetical core melt accident. For start-up and normal operation of the plant, a dedicated feed The tank is located at the bottom of the containment below the system, separate from EFWS, is provided. operating floor, between the reactor cavity and the radiological shield. In postulated accident conditions, the IRWST content is cooled by ESSENTIAL SERVICE WATER the LHSI system. The Essential Service Water System (ESWS) consists of four Screens are provided to protect the SIS, CHRS and CVCS pumps separate safety trains which cool the CCWS heat exchangers using from debris that might become entrained with IRWST fluid under water from the heat sink in all normal plant operating conditions and accident conditions. during incidents and accidents.
Emergency Feedwater System (EFWS)
– Interconnecting headers at EFWS MSRV MSSV MSIV pump suction and discharge 1 x 50 % 2 x 25 % normally closed. EFWP – Additional diverse electric power supply for 2/4 trains, using two small Diesel generator sets. MSRV MSSV MSIV 1 x 50 % 2 x 25 % EFWP To turbine
MSRV MSSV MSIV 1 x 50 % 2 x 25 % EFWP
MSRV : Main Steam Relief Valves. MSSV : Main Steam Safety Valves. MSIV : Main Steam Isolation Valves. MSRV MSSV MSIV 1 x 50 % 2 x 25 % EFWP
22 I ULTIMATE COOLING WATER SYSTEM COMPONENT COOLING WATER The Ultimate Cooling Water System (UCWS) is a diverse system The Component Cooling Water System (CCWS) transfers heat from allowing the dedicated cooling system associated with the mitigation the safety related systems, operational auxiliary systems and other of postulated severe accidents to be cooled, or to act as a back up reactor equipment, to the heat sink via the Essential Service Water for cooling the fuel pool. System (ESWS) under all normal operating conditions. The CCWS also performs the following safety functions: OTHER SAFETY SYSTEMS • it transfers heat from the SIS/RHRS to the ESWS, • it transfers heat from the Fuel Pool Cooling System (FPCS) to the The Extra Borating System (EBS) ensures sufficient boration of ESWS for as long as any fuel assemblies are located in the spent the RCS for transfer to the safe shutdown state at a boron fuel storage pool outside the containment, concentration required for cold shutdown. This system consists of • it cools the thermal barriers of the Reactor Coolant Pump (RCP) two separate and independent trains, each capable of injecting the seals, total amount of concentrated boric acid required to reach the cold • it transfers heat from the chillers in divisions 2 and 3 and cools the shutdown condition from any steady state power operating state. Containment Heat Removal System (CHRS) via two separate trains. The Main Steam System (MSS) upstream of the Main Steam The CCWS consists of four separate safety trains each located in Isolation Valves is safety classified. This part consists of four one of the four divisions of the Safeguard Buildings. geographically separated but identical trains, each including one main steam isolation valve, one main steam relief valve, one main steam relief isolation valve and two spring-loaded main steam safety OTHER SYSTEMS valves. Other systems include the Nuclear Sampling, Nuclear Island Vent The Main Feedwater System (MFS) upstream of the Main Feedwater and Drain, Steam Generator Blowdown, and Waste Treatment Isolation Valves is also safety classified. This consists of four Systems. geographically separated but identical trains, each including main • The Nuclear Sampling System is used for taking samples of gases feedwater isolation and control valves. and liquid from systems and equipment located inside the reactor In addition to the safety systems described above, other safety containment. functions are performed to mitigate postulated severe accidents, • The Vent and Drain System collects gaseous and liquid waste as described in the section dealing with safety and severe accidents. from systems and equipment for treatment. • The Steam Generator Blowdown System prevents build-up of solid matter in the secondary side water. • The Waste Treatment System ensures the treatment of solid, gaseous and liquid wastes.
BACK-UP FUNCTIONS AVAILABLE IN THE SAFETY SYSTEMS AND FUNCTIONS EVENT OF TOTAL LOSS OF THE REDUNDANT ➡ Simplification by separation of operating SAFETY SYSTEMS and safety functions. ➡ The loss of secondary side heat removal ➡ Fourfold redundancy is applied to the is backed up by primary side feed and safety systems and to their support bleed through an appropriately designed systems whenever maintenance during and qualified primary system overpressure operation is desired. This architecture protection system. allows them to be maintained during plant ➡ A combined function comprising operation, ensuring a high plant availability secondary side heat removal, accumulator factor. injection and the operation of the LHSI ➡ The different trains of the safeguard systems Systems can replace the MHSI System are located in four different buildings to function in the event of a small break loss which strict physical separation is applied. of coolant accident. ➡ Because of the systematic application ➡ Similarly, complete loss of the LHSI system of functional diversity, there is always is backed up by the MHSI system and a diverse system which can perform the by the Containment Heat Removal System desired function and bring the plant back (CHRS) for IRWST cooling. to a safe condition in the highly unlikely event of all the redundant trains of a system becoming simultaneously unavailable.
I 23 ■ EPR™ NUCLEAR ISLAND
POWER SUPPLY The outline design of the power supply system is shown below. The Emergency Power Supply is designed to ensure that the safety systems are supplied with electrical power in the event of loss of the preferred electrical sources. The Emergency Power Supply comprises four separate and redundant trains arranged in accordance with the four division concept. Each train is provided with an Emergency Diesel Generator (EDG) set. The emergency power supply system is designed to meet the requirements of the N+2 concept. The safety loads connected to the emergency power supply are those required to safely shut down the reactor, remove residual and stored heat, and prevent the release of radioactivity. In the event of total loss of the four EDGs (Station Black Out – SBO), two additional generators, the SBO Diesel Generators, provide the power necessary to supply the emergency loads. The SBO Emergency Diesel Generators are connected to the safety busbars of two divisions. Isar-2, Germany (KONVOI, 1,300 MWe) emergency Diesel generator.
Electrical systems of an EPR™ nuclear power plant (typical)
Stand-by grid Main including auxiliary grid stand-by transformer
Auxiliary Auxiliary normal normal transformer transformer G
10kV 10kV 10kV 10kV
M M M M
690V 690V 690V 690V 400V 400V 400V 400V
Turbine Island
Nuclear Island EDG EDG EDG EDG RCP RCP RCP RCP G M G M G M G M 10kV 10kV 10kV 10kV
M M M M G G SBO 690V 690V 690V 690V SBO 400V 400V 400V 400V
24 I FUEL HANDLING AND STORAGE and is designed to keep the SFP temperature at the required level during normal plant operation (power operation and in refuelling The reactor core is periodically reloaded with fresh fuel assemblies. outages). This system is arranged in two separate and independent The spent fuel assemblies are moved to and stored in the Spent Fuel trains with two FPCS pumps operating in parallel in each train. Pool (SFP). These operations are carried out using several handling The FPCS also includes a third diverse cooling train to cope with devices and systems (fuel transfer tube, spent fuel crane, fuel elevator, any common mode failure of the two main trains (including loss of the refuelling machine and spent fuel cask transfer machine). CCWS). The underwater fuel storage racks are used for underwater storage of: The FPPS comprises a purification loop for the SFP, a purification • fresh fuel assemblies, prior to loading, loop for the reactor pool and the IRWST, and skimming loops for • spent fuel assemblies following fuel unloading from the core and the SFP and the reactor pool. The system includes two cartridge prior to shipment off the plant. filters, a demineralizer and a resin trap filter used for purification The Fuel Pool Cooling and Purification System (FPCPS) is divided of pool water. into two subsystems: the Fuel Pool Cooling System (FPCS) and the Fuel Pool Purification System (FPPS). The FPCS provides the capability for heat removal from the SFP
Chooz B1, France (N4, 1,500 MWe) Fuel Building.
I 25 ■ EPR™ NUCLEAR ISLAND
INSTRUMENTATION & CONTROL SYSTEM
A nuclear power plant, like any other industrial facility, requires means for monitoring and controlling its processes and equipment. These means, as a whole, constitute the plant Instrumentation & Control (I&C) system, which comprises several subsystems with their electrical and electronic equipment.
The I&C system is basically composed of sensors to transform Safety classification physical data into electrical signals, programmable controllers to process these signals and provide actuator control and I&C functions and equipment are categorized into classes depending monitoring and control means for use by the plant operators. on their importance to safety. I&C functions are implemented using components with the appropriate quality level for their safety class. The overall design of the I&C system and associated equipment must comply with process, nuclear safety, and operational Redundancy, separation, diversity requirements. and reliability ➡A computerized plant I&C system, EPR™ I&C systems and equipment comply with the principles of supported by modern digital technologies. redundancy, diversity and separation applied in the design of EPR™ safety-related systems. For example the Safety Injection System and the Emergency Feedwater System, which each consist of four EPR™ I&C OVERALL ARCHITECTURE redundant and independent trains, also have four redundant and Within the overall I&C architecture, each subsystem is characterized independent I&C channels. according to its functions (measurement, actuation, automation, man- Each safety-related I&C system is designed to be able to machine interface) and its role in safety or operation of the plant. satisfactorily fulfill its function, even if one of its channels is not available due to a failure and a second one is unavailable for A multi-level structure preventive maintenance reasons. The structure of the EPR™ Instrumentation and Control System is The level of availability of the I&C systems performing safety functions characterized by a four-level organisation: is specified so as to comply with the probabilistic safety targets • Level 0 corresponds to sensors and actuators; adopted in the EPR™ design. • Level 1 carries out the automation functions: it comprises the reactor control and protection systems, the turbo-generator control ➡A quadruple redundant safety-related I&C and protection system, and the system performing all other for a further increase in the level of safety. automation functions (plant control and protection); • Level 2 carries out functions related to the human-machine interface that allow the plant to be operated and monitored; • Level 3 contains computer applications designed for non-real- time operation (dedicated to so-called “operating” personnel, i.e. staff carrying out operation and maintenance) and applications used outside of the plant. These applications provide operational and maintenance aids (trends, reports…), which assist in off-line analysis of the state of the unit or equipment and interlock management. They also permit interfacing and exchange of data with offsite bodies (power distribution network dispatching, National Emergency Support Centers, etc.). The architecture is supported by computer aided design (CAD) and a data production organisation, which ensures the consistency of the configuration (Design CAD is used for functional studies and data production CAD for the Instrumentation and Control systems programming).
26 I A computerized screen-based control room designed to maximize operator efficiency. Description of the I&C architecture Chooz B1, France (N4, 1,500 MWe).
Functional safety class* F1A Functions required in case of accident to bring the reactor to controlled state. F1B Functions required after an accident to bring the reactor to safe state. Functions intended to avoid the risk of radioactive releases. F2 Other functions contributing to plant safety (adherence to limit operating conditions, surveillance of safety system availability, protection against the effects of internally- generated hazards, detection/monitoring of radioactive releases, functions used in post-accident operation, etc.). NC Non-classified functions.
* Defines Quality Assurance Requirements.
I&C architecture (functional diagram)
Remote Main control room Maintenance Technical shutdown station technical room support center
SICS
SICS Safety Information PICS & Control System PICS Process Information & Control System Reactor Control, Specific I&C RCSL Surveillance and PAS RCSL PS SAS Limitation System (TPCS, etc.) PS Protection System Safety Automation SAS System Process Automation PAS System PACS Priority and Actuator PACS Control System Control Rod Drive CRDM Mechanism TPCS Turbine Protection & Control System instrumentation actuators control trip instrumentation actuators Diversified TXS* Operational systems CRDM Safety systems technology
* TELEPERM-XS AREVA NP technology. XXX: safety class F1A and F1B.
I 27 ■EPR™ NUCLEAR ISLAND
ROLE OF THE I&C SYSTEMS The SAS (Safety Automation System), the functions of which are: Like all other EPR™ systems, the I&C systems act in accordance • managing F1B post-accident processing to take the unit from the with the “defense in depth” concept (see page 32). controlled state to the safe state; Three lines of defense are implemented: • preventing significant radioactive releases (F1B); • the control system maintains plant parameters within their normal • management of events which could lead to accidents or limiting operating ranges, accidents. • in case a parameter leaves its normal range, the limitation system The PS (Protection System) performs the following functions: performs appropriate actions to avoid the need for the initiation of • monitoring of safety parameters in all unit operating conditions; protective actions, • enabling, following an initiating event: • if a parameter exceeds a protection threshold, the reactor protection – automatic F1A protection and safeguard actions, system generates the appropriate safety actions (reactor trip and – automatic F1A commands of the safety support systems; safety system actuation). • possible manual F1A actions. Normally, to operate and monitor the plant, the operators use Diversity of equipment (software and hardware) between workstations and a plant overview panel in the Main Control Room. the PAS and the PS is implemented in order to reduce the In case of unavailability of the Main Control Room, the plant is risk of a common mode failure. monitored and controlled from the Remote Shutdown Station. The RCSL (Reactor Control Surveillance and Limitation) Instrumentation carries out: • monitoring of the core and the rod control function; A number of instrumentation channels supply measured data for • monitoring of the core and the reactor coolant system against the control systems, surveillance and protection systems, and for limiting conditions of operation; information for the control room staff. Multiple-channel acquisition • limitation actions to anticipate and avoid the possibility of a reactor is used for control of important parameters such as the pressure protection threshold being exceeded. and temperature of the primary coolant and the liquid level in the reactor pressure vessel. Multiple-channel and diverse data The PACS (Priority and Actuator Control System) manages acquisition methods are implemented. the control and monitoring of the actuators used by both the operational and the safety systems. For actuators involved in Role of the various I&C subsystems protection (F1A) or post-accident functions (F1B), the assignment of priorities to commands from the PS (F1A), SAS (F1B) and/or The automation system can be broken down into two subsystems: PAS (NC/F2) is managed by the PACS. The PAS (Process Automation System), which carries out: The PACS is responsible for four types of functions: • automation and monitoring of the unit in normal operating • priority management; conditions. The PAS processes F2 and non-classified (NC) • essential protection of components; functions with the exception of those carried out by the RCSL • actuator control; (defined below) and the control and protection system for the • actuator monitoring. conventional island and site-dedicated equipment; Functions are allocated and distributed over several pieces of • the PAS is used for the monitoring and automation of F2 and NC equipment (switchgear panels and Level 1 equipment). functions to mitigate the consequences of severe accidents and multiple failure events (risk reduction category events); All the I&C subsystems are implemented using digital • limitation actions to anticipate and avoid an eventual trip of the equipment. reactor due to a protection system threshold being exceeded.
TELEPERM XS PS, RCSL and PACS are implemented using the TELEPERM XS and standards (IEC, IEEE, EPRI and KTA). It is complemented technology. TELEPERM XS is AREVA NP’s digital I&C platform by engineering tools and associated equipment supporting specifically developed for safety and high-reliability functions in all design phases. nuclear facilities. Its high functional reliability is achieved thanks In May 2000, the US Nuclear Regulatory Commission issued to a combination of fail-safe design, fault tolerance, integrated the generic approval for the use of the TELEPERM XS system self-checking, structural simplicity and outstanding robustness. platform in all safety applications, including protection systems. Its resistance to environmental conditions such as temperature TELEPERM XS is also licensed and used in Argentina, Bulgaria, swings, vibrations, seismic loadings and electromagnetic China, Finland, France, Germany, Hungary, Slovakia, Spain, radiations meets the requirements of international codes Sweden and Switzerland.
28 I Human-machine interface (level 2) The architecture of the EPR™ Instrumentation and Control System is evolutionary with regard The EPR™ Control Room comprises: to existing plants • a computerized Main Control System which allows the unit The main developments have been in the following areas: to be run in normal and accident situations. It comprises: – four operator workstations with five standardized operating screens • drawing on the lessons of the N4 series and the German KONVOI and a dedicated screen for the operating applications (level 3), plants (architecture in four divisions, operator stations with five – several Compact Operator Workplaces (COW), i.e., workplaces standardized screens fitted with a signalling strip for alarms, with a reduced number of screens which can be configured in a automatic limitation functions…); control mode or in supervision mode and used for maintenance • increasing safety and availability of the unit (architecture in and outage phases; four divisions, automation of plant limits…); a mimic panel in the control room which provides a common • • optimizing performance and costs, whilst still being based on reference for the various members of the operator team; mature technological developments (proven Teleperm XS • a Safety Information and Control System (SICS) to be used components, computerized mimic panel…). if the Process information and Control System (PICS) is not available. It allows the unit to be brought to, and kept in, a safe shutdown state. In order to reduce the risk of a common mode failure, the items of equipment used for the PICS and the SICS are independent. In the event of an unavailibility of the control room (e.g. fire), the Remote Shutdown Station (RSS), which is located outside the control room, allows operators to bring the unit to a safe state. The RSS will be given computerized control means with technology which is identical to that of the PICS. The RSS comprises three operator stations.
The EPR™ power plant's computerized control room features control screens providing relevant summary information on the process (computer-generated picture).
I 29 SAFETY
> NUCLEAR SAFETY page 31
THREE PROTECTIVE BARRIERS page 31
DEFENSE IN DEPTH page 32
> EPR™ SAFETY SYSTEMS page 33
DESIGN CHOICES FOR REDUCING THE PROBABILITY OF ACCIDENTS LIABLE TO CAUSE CORE MELT page 33 Golfech 2, France (1,300 MWe): reactor pressure vessel and internals. DESIGN CHOICES FOR LIMITING THE CONSEQUENCES OF SEVERE ACCIDENTS page 36
30 I NUCLEAR SAFETY
The fission of atomic nuclei, which takes place in a nuclear reactor to generate heat, also produces radioactive substances from which people and the environment must be protected. Nuclear safety is the set of technical and organisational provisions that are applied in the design, construction and operation of a nuclear plant to reduce the likelihood of an accident and to limit its consequences should it nevertheless occur.
Nuclear reactor safety requires that at all times three basic THREE PROTECTIVE BARRIERS safety functions should be fulfilled: • control of the nuclear chain reaction, and therefore of the power The concept of the “protective barriers” involves placing a series of generated, strong, leak-tight physical barriers between the radioactive materials • cooling of the fuel, including removal of residual heat after the chain and the environment to contain radioactivity in all circumstances: first barrier: reaction has stopped, • the fuel, inside which most of the radioactive products • containment of radioactive products. are already trapped, is enclosed within a metal cladding; • second barrier: the reactor coolant system is enclosed within a Nuclear safety relies upon two main principles: pressurized metal envelope that includes the reactor vessel which • the availability of three protective barriers, houses the core containing the fuel rods; • application of defense in depth. • third barrier: the reactor coolant system is itself enclosed in a thick- walled concrete containment building (for the EPR™ reactor, the containment is a double shell resting on a thick basemat, the inner The three protective barriers wall being covered with a leak-tight metallic liner).
➡ Maintaining the integrity and leaktightness of just one of these barriers is sufficient to contain radioactive fission products. Steam generator
Pressurizer
Control rod drive Reactor mechanism coolant pump
3 2 1 Fuel assembly Reactor pressure vessel
1 Fuel cladding 2 Reactor coolant boundary 3 Reactor containment
I 31 ■SAFETY
DEFENSE IN DEPTH also have redundancy, are provided to ensure containment of radioactive products; The concept of “defense in depth” involves ensuring the • to further extend the defense in depth approach a failure of all effectiveness of the protective barriers by identifying the threats to three levels is postulated, resulting in a “severe accident” situation. their integrity and by providing successive lines of defence to protect Means are provided to minimise the consequences of such a them from failure: situation. • first level: implementation of a safe design, high quality of construction and safe and reliable operation incorporating lessons ➡ Applying the defense in depth concept from experience feedback in order to prevent occurrence of failures; leads to the functions of core power and • second level: means of surveillance for detecting anomalies that cooling control being protected by multiple could lead to a departure from normal operating conditions, in order redundant systems: fourfold redundancy to anticipate failures or to detect and intercept deviations from is used in the EPR™ technology. normal operational states in order to prevent anticipated operational ➡ Safety functions are ensured by diversified occurences from escalading to accident conditions. The most means to minimize the risk of common important of this level is the one that automatically shuts down the mode failure. reactor by insertion of the control rods into the core which stops the nuclear chain reaction in a few seconds; ➡ In addition, the components of these • third level: means of action for mitigating the consequences systems are designed to automatically of failures and preventing core melt down. This level includes use move to a safe position in case of a failure of diverse and redundant systems to automatically bring the reactor or a loss of electrical or power supplies. to a safe shutdown state. In addition, a set of safety systems, which
The training for steam generator inspection illustrates: ➡ the first level of defence in depth relating to the quality of workmanship, ➡ the second barrier, as the training relates to steam generator tubes which form part of the primary system.
Lynchburg technical center (Va, USA): training for steam generator inspection.
32 I EPR™ SAFETY SYSTEMS
A key decision, in line with the recommendations of the French and German safety authorities, was to base the EPR™ design on an evolutionary approach using experience feedback from more than 100 reactors previously built or under construction by AREVA NP. This decision enabled the designers to use experience from the most recently constructed plants (N4 reactors in France and KONVOI in Germany) and to avoid the risk arising from the adoption of unproven technologies. This approach did not mean that innovative solutions, backed by the results of large-scale Research and Development programs, were excluded. Key innovations have been included in the EPR™ design to help accomplish EPR™ safety systems objectives, in particular with regard to the prevention and mitigation of hypothetical severe (core melt) accidents.
The EPR™ safety approach, motivated by a desire to steadily • an extended range of operating conditions was taken into account increase the level of safety, involves a reinforced application of the at the design stage, defense in depth concept: • equipment and systems were designed to reduce the likelihood of • by improving preventive measures to reduce the probability of core an abnormal situation deteriorating into a severe accident, melt, • improvements were made in the reliability of operator actions. • by incorporating features for limiting the consequences of core melt accidents at the design stage. Extension of the range of operating conditions considered at design stage ➡ A twofold safety approach is used against severe accidents: The use of the probabilistic safety assessments • reducing their probability by reinforced Although the EPR™ safety approach is based mainly on applying the preventive measures, defense in depth concept (which is part of a deterministic approach), • drastically limiting their potential the design is supported by probabilistic analyses. These make consequences. it possible to identify accident sequences that could cause core melt or result in large radioactivity releases, to evaluate their probability, and to identify their potential causes and countermeasures. The use DESIGN CHOICES FOR REDUCING THE of probabilistic assessment at the design phase of the EPR™ reactor PROBABILITY OF ACCIDENTS THAT COULD has been a decisive factor in the choice of technical options to improve CAUSE CORE MELT the safety level of the reactor. In order to reduce the probability of core melt accidents, below the For EPR™ technology, the probability of an accident leading to core melt already extremely low levels achieved in reactors in the French and is extremely small and below that in the previous-generation reactors: German nuclear power plant fleet, advances were made in three • below 1/100,000 (10–5) per reactor/year for all types of initiating failure areas: and hazard, which meets the objective set for new nuclear power plants established by the International Nuclear Safety Advisory Group (INSAG) with the International Atomic Energy Agency (IAEA) – INSAG 3 report, The EPR™ technology complies with the • below 1/1,000,000 (10–6) per reactor/year for events occurring inside safety objectives set down by the French the plant, which is a factor 10 reduction compared with the most and German safety authorities for future modern reactors currently in operation, PWR power plants: • below 1/10,000,000 (10–7) per reactor/year for the accident ➡ further reduction of core melt probability, sequences associated with early loss of the radioactive containment function. ➡ practical elimination of accident situations which could lead to a large Consideration of shutdown states in the design early release of radioactive materials, of protection and safety systems ➡ need for only very limited protective Probabilistic safety assessments highlighted the importance of measures in area and time*, in case reactor shutdown states. These shutdown states were systematically of a postulated low pressure core melt taken into account in the EPR™ design, both in risk analysis and in situation. the design of the protection and safety systems. * No permanent relocation, no need for emergency evacuation outside the immediate vicinity of the plant, limited sheltering, no long-term restriction in the consumption of food.
I 33 ■SAFETY
Greater protection from internal and external hazards The layout of the safety systems and the design of the civil works structures minimize the risks from hazards such as earthquake, flooding, fire, airplane crash. The safety systems are designed on the basis of a quadruple redundancy, both for their mechanical and electrical parts and for of 1 1 the supporting I&C. This means that each system consists of four subsystems, or “trains”, each one capable by itself of fulfilling the 1 entire safety function. The four redundant trains are physically separated from each other and located in four independent divisions (buildings). 2 Each division includes one train of: • the safety injection system for injecting borated water into the reactor vessel in a loss of coolant accident. This consists of a low-head injection system and its cooling loop, and a medium-head injection system, 1 • the steam generator emergency feedwater system, • the electrical and I&C systems supporting these systems. The building housing the reactor, the building in which the spent fuel is stored on an interim basis, and the four buildings corresponding to the four divisions of the safety system are provided with special protection against externally generated hazards such as earthquakes and explosions. The major safety systems comprise four sub-systems or trains, each capable of performing the entire safety function on its own. There is one train in each Protection against an aircraft crash has been further strengthened. of the four safeguard buildings (1) surrounding the reactor building (2) to prevent The reactor building is protected by a double concrete shell: an outer a simultaneous failure of the trains. thick shell made of reinforced concrete and an inner thick shell made of pre-stressed concrete which is internally covered with a thick metallic liner. The thickness and the reinforcement of the outer shell ➡A set of quadruply redundant safety provide sufficient strength to absorb the impact of a large commercial systems, with independent and aircraft. The double concrete wall is extended to the fuel building, geographically separated trains, and to two of the four safeguard buildings containing the Main minimize the consequences of potential Control Room and the remote shutdown station which would be internal and external hazards. used in emergency conditions. ➡Protection against aircraft crash is The other two safeguard buildings which are not protected by the reinforced by use of a strong double- double wall are remote from each other and separated by the reactor walled concrete shell implemented building, which prevents them from being simultaneously damaged. to shelter the EPR™ reactor. In this way, if an aircraft crash were to occur, at least three of the four trains of the safety systems would be available.
Choice of equipment and systems to reduce The outer shell (5) covers the the risk of an abnormal situation deteriorating reactor building (2), the spent into an accident fuel building (3) and two of the four safeguard buildings (1). Reduction in the risk of a large reactor coolant pipe break 1 The other two safeguard buildings are separated 1 The design of the reactor coolant system, the use of forged pipework 2 geographically. and components, construction with high mechanical performance materials, combined with the measures to allow early leak detection 5 and to facilitate in-service inspections, allow rupture of the major 3 reactor coolant pipework to be excluded from the design basis 4 (i.e. to be “practically eliminated” by design).
The reactor containment building has two walls: an inner pre-stressed concrete housing (4) internally covered with a metallic liner and an outer reinforced concrete shell (5).
34 I Optimized management of postulated steam further equipped with a dedicated system for cooling the reactor generator tube break containment in severe accidents which would be activated by an operator only in the event of a core melt accident. Steam generator tube break could potentially result in a transfer of water from the primary system to the secondary system. Following Residual heat removal is provided by the four trains of the low head such an event the fall in the primary side pressure would part of the safety injection system, which under these circumstances automatically induce a reactor shutdown and then activate safety would be configured to remove the residual heat in closed loop mode injection into the reactor vessel. In the EPR™ reactor, the driving (suction via the hot legs, discharge into the cold legs). Safety pressure of the medium-head injection is set below the set pressure injection remains available for action in the event of a leak or break of the secondary system safety valves, which prevents the steam occurring on the reactor coolant system. generators from overfilling with water in these circumstances. This has the advantages of avoiding the release of liquid coolant, thus ➡ The safety-related systems are simple, reducing the potential release of radioactivity into the environment, redundant and diverse to ensure high and also considerably reducing the risk of a secondary system safety reliability and effectiveness. valve jamming in open position.
Simplification of the safety systems and optimization Increased reliability of operator action of their redundancy and diversity Extension of action times available to the operator Those safety systems with a quadruple redundancy are spread in four separate divisions. The short term protection and safety actions needed in the event of an incident or accident are automated. To ensure a high level The system design is straightforward with minimal changes being of safety, design criteria have been established to set minimum required to the system configuration whether the reactor is at power timeframes before operator action is required. In any case, operator or shutdown. This is illustrated by the design of the EPR™ safety action is not required before at least thirty minutes for actions taken injection system and residual heat removal system. in the Control Room, or one hour for actions performed locally on The safety injection system, which would be activated in a loss of the plant. coolant accident, is designed to inject water into the reactor circuit The increased volume of the major EPR™ components (reactor to cool the reactor core. In the first phase of injection, water is pressure vessel, steam generators, pressurizer) increases the injected into the cold legs of the reactor coolant system loops response time of the reactor in upset conditions, extending the (pipework sections located between the reactor coolant pumps and timescales available to the operators to carry out initial actions. the reactor vessel). In the longer term, water is simultaneously injected into the cold and Increased performance of the human-machine interface hot legs (legs located between the steam generators and the reactor Experience feedback from the design and operation of the N4 reactors, vessel). The requirement for switching from the so-called “direct which were among the first plants to be equipped with a fully injection” phase to a “recirculation” phase in previous reactor computerized Control Room, and use of the last generation, yet well designs, does not apply to the EPR™ reactor. The EPR™ low head proven Teleperm XS safety I&C give the EPR™ reactor a high safety injection system is provided with heat exchangers enabling it performance and reliable human-machine interface. Operator actions carry out the core cooling function on its own. The EPR™ reactor is are based on real time plant data made available by the state-of-the- art EPR™ I&C.
➡ Design of components, high degree of automation, advanced design of I&C and human-machine interface combine to increase the reliability of operator actions.
The ergonomics of the EPR™ Control Room benefits from the latest developments in human-machine interface design (computer-generated picture).
I 35 ■SAFETY
DESIGN CHOICES FOR LIMITING THE Prevention of high-energy CONSEQUENCES OF SEVERE ACCIDENTS corium/water interaction
➡ A core melt-accident, in itself highly The high mechanical strength of the reactor vessel prevents it from unlikely, would only require very limited being significantly damaged by any conceivable reaction, which off-site countermeasures both with regard could occur between corium* and coolant inside the vessel. to time and the extent of the affected area. The areas of the containment where the corium could come into contact with water after being ejected from the pressure vessel – In response to the new safety requirements for future nuclear power namely the reactor pit and the core spreading area – are kept “dry” plants, introduced as early as 1993 by the French and German safety (free of water) in most circumstances. Only when the corium is authorities, the plant design must be such that a core melt accident spread inside the dedicated spreading area, then partially cooled would need only very limited off-site countermeasures in time and space. and solidified (and therefore less reactive), is it brought into contact The policy of mitigation of the consequences of a severe accident, with cooling water. which guided the design of the EPR™ reactor, therefore aimed to: ➡ “practically eliminate” situations which could lead to early Containment design against hydrogen risks radiological releases, such as: In the unlikely event of a severe accident, hydrogen could be released • high-pressure core melt ejection from the reactor pressure inside the containment in significant quantities. Hydrogen may be vessel, produced initially by reaction between the coolant and the zirconium • high-energy corium*/water interactions, used in the fuel rod cladding, and subsequently by the reaction • hydrogen detonations inside the reactor containment, between the corium and the concrete in the corium spreading and • by-pass of the containment; cooling area. ➡ ensure the integrity of the reactor containment, even in the event of a low-pressure core melt followed by ex-vessel For this reason, the pre-stressed concrete inner shell of the progression, through: containment is designed to withstand pressures that could • retention and stabilisation of the molten corium* inside the conceivably result from the combustion of this hydrogen. Further containment, devices called catalytic hydrogen recombiners are installed inside • cooling of the corium. the containment to keep the average concentration below 10% at all times, to avoid the risk of detonation. Assuming hydrogen combustion ➡ Situations which could generate by deflagration only, the pressure in the containment will not exceed a significant radioactivity release are 5.5 bar. “practically eliminated” by design. Corium retention and stabilization to protect Prevention of high-pressure core melt the basemat In addition to the reactor coolant system depressurisation systems The reactor pit is designed to collect the corium in the case of ex- provided on the other reactors, the EPR™ reactor is equipped with valves vessel melt progression and to transfer it to the corium spreading dedicated to preventing ejection of molten core materials at high-pressure and cooling area. The floor of the reactor pit is protected by in the event of a severe accident. These valves ensure fast depressurization, “sacrificial” concrete which is backed-up by a protective layer even in the event of failure of pressurizer relief lines. consisting of refractory concrete. The valves, which are controlled by the operator, are designed to safely remain in open position after their first actuation. Their high- relief capacity enables fast primary depressurisation of the reactor coolant system to a pressure of a few bars, precluding any risk of over-pressurization of the containment through dispersion of corium debris in the event of vessel failure at high pressure.
* Corium: product resulting from the melting of the core components and interaction with the structures it could meet.
➡ Even in the extremely unlikely event of a core melt accident resulting in ejection of molten materials from the reactor pressure vessel, the molten core and radioactive products would be most likely to remain confined inside the reactor building whose long term integrity would be In the event of core meltdown, molten core escaping from the reactor vessel would be passively maintained. collected and retained, then cooled in a specific area inside the reactor building.
36 I The dedicated corium spreading and cooling area (core-catcher) period is available for the deployment of this system by the operators consists of a channelled metal structure covered with “sacrificial” (at least 12 hours) while the large containment free volume helps to concrete. Its purpose is to protect the nuclear island basemat from limit the pressure increase after the accident. damage. Its lower section contains cooling channels in which water A second mode of operation of the containment heat removal system is circulated. The large spreading surface area (170 m2) promotes enables it to feed water directly into the core-catcher to cool it. cooling of the corium. The transfer of the corium from the reactor pit to the spreading area Collection of inter-containment leaks is initiated by a passive device: a steel “plug” that melts due to the heat from the corium. In the event of a core melt leading to vessel failure, the containment building would be the only remaining barrier of the three containment After spreading, flooding of the corium is triggered by opening of a barriers; provisions are therefore made to ensure that it remains passively activated valve. It is then cooled, also passively, by gravity undamaged and leak-tight. For the EPR™ technology, the following injection of water from the refuelling water storage tank located inside measures are adopted: the containment and evaporation. • a metal liner internally covers the pre-stressed concrete inner shell, The cooling stabilises the corium in a few hours and ensures its • the internal containment penetrations are equipped with redundant complete solidification in a few days. isolation valves and leak recovery devices to avoid containment bypass, Containment heat removal system • the architecture of the peripheral buildings and the sealing systems of the penetrations removes the risk of direct leakage from the inner and long-term residual heat removal device containment to the environment, In order to maintain the long-term integrity of the containment in a • the inter-space between the inner and outer shells of the severe accident means are provided to control the pressure increase containment is maintained at a slightly negative pressure to enable inside the containment due to residual heat. A dedicated dual-train leaks from the inner containment to be collected, spray system with heat-exchangers and dedicated heat sink is • the above provisions are supplemented by a containment ventilation provided to fulfil this cooling function. Owing to the thermal inertia of system and a filter system upstream of the stack. the containment and its internal concrete structures, a long time
Containment heat removal Containment spray system system
Spray nozzles Water injection Recirculation and into the core coolant heat exchanger catcher
Passive x x flooding device
x
CHRS Corium In-containment refueling (2x) spreading area water storage tank
Melt flooding via cooling device x Water level in case of water and lateral gap injection into spreading area
FL Flow limiter
Two fully redundant trains with specific diversified heat sink
I 37 EPR™ REACTOR CONSTRUCTION
> EPR™ REACTOR CONSTRUCTION TIME SCHEDULE page 39
DESIGN FEATURES page 39
CONSTRUCTION AND INSTALLATION METHODS page 39 Olkiluoto 3 nuclear power plant, Finland June 2009. MAJOR COMPONENT MANUFACTURING page 39 COMMISSIONING TESTS page 39
38 I EPR™ REACTOR CONSTRUCTION TIME SCHEDULE
The evolutionary approach adopted for the EPR™ design enables its construction schedule to benefit from the extensive construction experience feedback and the continuous improvements in processes, methodologies and tasks sequencing achieved by AREVA NP worldwide. Provisions have been made in the design, construction, and commissioning methods used to shorten the EPR™ reactor construction schedule as far as possible. Significant examples are as follows.
DESIGN FEATURES COMMISSIONING TESTS The general design layout of the main safety systems in four trains As with the interfaces between civil and construction works, the housed in four separate buildings allows parallel installation works interfaces between construction and testing have been carefully and, later, parallel testing of individual systems. reviewed and optimized. For instance, teams responsible for the commissioning tests are also involved in the finishing operations, Location of electromechanical equipment lower levels means that flushing, and system conformance checks, so that these activities its installation can start early and consequently remove it from critical are only carried out once. path, which contributes to the shortening of the construction schedule. Instrumentation & Control factory acceptance tests are carried out on a single test platform with all cabinets interconnected, which ensures a shorter on-site test period together with improved overall CONSTRUCTION AND INSTALLATION METHODS quality. Three main principles are applied to EPR™ reactor construction and The benefits drawn from the unique experience feedback gained by installation: minimization of the interfaces between civil works and installation AREVA NP in previous projects, and systematic optimisation of of mechanical components, modularisation and piping prefabrication. construction and testing activities and their interfaces, results in an Minimization of the interfaces between civil works and installation. optimal technical and economical construction schedule for The continuing search for optimisation of the interfaces between civil EPR™ reactor implementation. This experience and that from current and installation works results in the implementation of a construction EPR™ reactor projects, gives confidence that the planned EPR™ methodology “per level” or “grouped levels”. This enables equipment reactor time schedule is realistic. and system installation work at level “N”, finishing construction work at level “N+1” and main construction work at levels “N+2” and “N+3” to be carried out simultaneously. The methodology is used for all the different buildings except for the Reactor Building, where it cannot apply. Indicative planning and overall time-scale Use of modularization for overall schedule optimization. Modularization The overall construction schedule of a unit in the series depends techniques are considered systematically, but retained only in cases where largely on site conditions, industrial organisation and policies, and they offer a real benefit to the optimization of the overall construction local working conditions. Therefore figures quoted for a specific schedule without introducing a technical and financial burden due to project cannot usually be extended to others. advanced detailed design, procurement or prefabrication. This approach enables the site preparation schedule to be optimized, delays investment costs with regard to start of operation, and so offers financial savings. EPR™ REACTOR TIME SCHEDULE For instance, modules are implemented for the civil works associated with the reactor building (e.g. containment liner & dome), pool structure Year 0 Year 1 Year 2 Year 3 Year 4 for vertical walls in Reactor & Fuel buildings, control room, as they are on Main contract the critical path. First concrete pouring Maximization of piping and support prefabrication. Piping and sup- Start fuel loading port prefabrication is maximized in order to minimize installation man-hours and especially welding and controls at installation locations; this measure Commercial operations also results in an even better quality of the piping spools at lower cost. Engineering
Manufacturing MAJOR COMPONENT MANUFACTURING Licensing AREVA NP’s Chalon/Saint-Marcel and Jeumont plants have over thirty years of experience in the manufacture of heavy nuclear Civil works components. This has resulted in considerable experience in optimization of the production time of heavy nuclear components. Installation The construction of the EPR™ reactor benefits from this unique Start-up tests manufacturing capability and expertise.
I 39 PLANT OPERATION, MAINTENANCE & SERVICES
AN AVAILABILITY DESIGN TARGET ABOVE 92% page 41
A HIGH LEVEL OF OPERATIONAL MANOEUVRABILITY page 42
Neckarwestheim nuclear power plant (Germany): AN ENHANCED RADIOLOGICAL unit 2 (right foreground) is of the KONVOI type PROTECTION page 42 (1,300 MWe). PLANT SERVICES page 43
40 I PLANT OPERATION, MAINTENANCE & SERVICES
Plant operators worldwide are trying hard to increase plant availability and to reduce maintenance costs. The EPR™ systems and components have been designed from the beginning to help them achieve these goals through efficient refueling outages and simplified inspection and maintenance.
AN AVAILABILITY DESIGN TARGET also facilitates field services which could be needed outside ABOVE 92% scheduled outage periods. Based on experience feedback, easier access to the components of the reactor allow simple and rapid The EPR™ reactor is designed to exceed an availability rate of 92%. performance of inspection and maintenance work. This is made possible by long cycles, short-scheduled outages for fuel loading/unloading and in-service inspections and maintenance, Access to the reactor building during power operation allows to start and also through reduced downtimes attributable to unscheduled preventive maintenance and outage preparation up to seven days outages. before reactor shutdown and to continue their demobilization up to three days after reactor restart. The quadruple redundancy of the safety systems allows a part of the preventive maintenance operations to be performed while the reactor The duration of the plant shutdown phase is reduced by a time gain is at power. for reactor coolant system cooldown, depressurization and vessel head opening. Similarly the length of the restart phase is reduced Moreover, the reactor building is designed to keep some areas as well and benefits from the reduction in the time needed to run accessible, under standard safety and radiation protection the beginning-of-cycle core physics tests (gain supplied by the conditions, while the reactor is at power. This enables the outage “aeroball” in-core instrumentation system). Durations of about and maintenance operations to be prepared and demobilized with no 70 and 90 hours are respectively scheduled for the shutdown and loss of availability. This possibility of access with the reactor on line restart phases. For the fuel loading/unloading operations, a time period below 100 hours is scheduled.
➡ Typical outage duration: the duration of a regular outage for preventive maintenance and refueling is reduced to 16 days. Duration of an outage for refueling only does not exceed 11 days. Decennial outages for main equipment in-service inspection, turbine overhaul and containment pressure test are planned to last 40 days.
The EPR™ reactor is designed to: ➡ maximize plant availability and manoeuvrability, ➡ ease operation and maintenance and reduce their costs, ➡ enhance radiological protection of the personnel, Chooz B1, France (N4, 1,500 MWe): removal of the hydraulic section of a reactor coolant ➡ protect the environment and contribute pump for maintenance. to a sustainable development.
41 I I 41 - PLANT OPERATION, MAINTENANCE & SERVICES
A HIGH LEVEL OF OPERATIONAL objective as soon as EPR™ design got underway. For this purpose, MANOEUVRABILITY the designers drew heavily upon the experience feedback from the operation of the French and German nuclear power plant fleets. In terms of operation, the EPR™ reactor is designed to offer the utilities a high level of manoeuvrability. It has the capacity to Accordingly, major progress has been made, particularly in the be permanently operated at any power level between 25 and following areas: 100% of its nominal power in a fully automatic way, with the • the choice of materials, for example the optimization of the quantity primary and secondary frequency controls in operation. and location of the Cobalt-containing materials and liners, in order to obtain a gain on the Cobalt 60 “source term”, The EPR™ reactor capability regarding manoeuvrability is a • the choices regarding the design and layout of the components particularly well adapted response to scheduled and unscheduled and systems liable to convey radioactivity, taking into account the power grid demands for load variations, managing of grid various plant operating states, perturbations or mitigation of grid failures. • the optimization of the radiation shielding thicknesses in response to forecast reactor maintenance during outages or in service. Thanks to these significant advances and to shorter outages, AN ENHANCED RADIOLOGICAL collective doses less than 0.4 Man.Sievert per reactor/year can PROTECTION be expected for operation and maintenance staff (to date, for the Allowance for operating constraints and for the human factor, with the major nuclear power plant fleets of OECD countries like France, aim of improving worker radiation protection and limiting radioactive Germany, the United States and Japan, the average collective dose releases, together with radwaste quantity and activity, was a set observed is about 1 Man.Sievert per reactor/year).
In-service inspection machine for ultrasonic testing of reactor pressure vessels.
42 I PLANT SERVICES The “FROG” Owners Group (see page below) offers member electricity companies a cost-effective means for exchange of information and Optimization of plant processes and implementation of innovative experience. FROG’s members have access to broad operational and maintenance technologies and concepts are also significant maintenance feedback. They also benefit from the results of study contributors to the achieving of operators’ cost and availability programs jointly decided to deal with issues of shared interest. objectives. In this area, AREVA NP supplies the most comprehensive range of nuclear services and technologies in the world. Thanks to its experience from designing and constructing over 100 nuclear power plants worldwide, its global network of maintenance and services centers with highly trained teams (more than 4,000 specialists mainly based in France, Germany and the USA) committed to excellence, AREVA NP provides a full range of inspection, repair and maintenance services for all types of nuclear power plants, based on the most advanced techniques available today. Its field of expertise covers the whole scope of customers’ needs from unique one-of-a- kind assignments to the implementation of integrated service packages with a view to reducing outage duration. In addition, AREVA is introducing innovative partnership models featuring long-term performance-based contracts for services. AREVA NP’s offer of power plant services encompasses: • in-service inspection and non destructive testing, • outage services, • asset management, • life cycle management, • predictive maintenance, • supply of spare parts, • off-site maintenance of components in “hot” workshops, • fuel inspection, repair and management, • services in the fields of instrumentation and diagnosis, I & C and electrical systems, chemistry, • plant decommissioning and waste management, • training of operating personnel, • expert consultancy.
THE “FROG” OWNERS GROUP
The FROG (formely Framatome Owners Group) British Energy owner of Sizewell B in the United Kingdom is dedicated to building strong and efficient teaming (in October 2002) joined the FROG as members. In 2003, for mutual cooperation, assistance and sharing GNPJVC and LANPC merged operation of their plants in of its members’ experience and expertise, to support one company DNMC. the safe, reliable, cost-effective operation of its members’ The Owners group provides a forum for its members to share nuclear power units. their experiences in all domains of nuclear power plant The FROG was set up in October 1991 by five utility operation, enabling a cost-effective exchange of information companies that were either operating or building nuclear to identify and solve common issues or problems. power plant units incorporating a Framatome nuclear Several working groups and technical committees are steam supply system or nuclear island. actively dealing with specific technical and management These utility companies are Electrabel from Belgium, issues. Among them, a specific Steam Generator Technical Electricité de France, Eskom from the Republic of South Committee, has been formed by utilities having steam Africa, GNPJVC from the People’s Republic of China generators served by AREVA NP. Committee participants and KHNP from the Republic of Korea. are the FROG members plus the companies NSP and Later on, Ringhals AB from Sweden (in June 1997), LANPC, AmerenUE from the USA, NOK from Switzerland and NEK owner of the Ling Ao plant in China (in October 2000), from Slovenia.
I 43 ENVIRONMENTAL IMPACT
> ENVIRONMENTAL IMPACT
DESIGN page 45
CONSTRUCTION page 45 Saint-Alban nuclear power plant, France (1,300 MWe x 2). OPERATIONS page 45
DECOMMISSIONING page 47
44 I ENVIRONMENTAL IMPACT
Ensuring that the impact of a nuclear unit on the environment is minimal requires complementary actions by the reactor vendor, the operator, the safety authorities and various other organizations. The vendor’s contribution to minimizing the environmental impact is set out below.
DESIGN CONSTRUCTION As already stated in the section on “EPR™ safety systems”, the During the building phase, activities such as clearance of the site, reactor has been designed to avoid the occurrence of incidents and excavation, drilling, concrete production and start-up tests have a accidents and to reduce their consequences for the plant and the potential impact on the environment. Therefore the marine and environment if they were to occur. terrestrial environments, freshwater, air, climate, landscape and the noise level must be kept under close scrutiny in order to minimize The improved thermodynamic efficiency of the plant ensures that its possible consequences. cooling water requirements are minimized. The high performance core of the EPR™ reactor (see section on the reactor core) gives the operator increased flexibility in the utilization OPERATIONS of nuclear fuel to generate power. Depending on the fuel The impact of operations on the environment translates into the management strategy adopted and on the relevant point of gaseous and liquid releases, and solid wastes produced. The comparison, the savings on uranium consumption per unit of energy evolutionary character of the EPR™ design makes it possible to draw produced can reach 7 to 15%, giving a corresponding reduction in benefit from lessons learned from many years of operation of earlier the quantities of high-level waste. generation reactors, meaning that whenever possible, releases and Processes have been optimized to reduce quantities of radioactive waste are reduced and, when this is not possible, the extent and and chemical waste through the use of the best tried and tested impact of such releases can be accurately predicted. methods available at an acceptable cost. Gaseous radioactive waste from an EPR™ reactor arises from: • the ventilation of the nuclear buildings; • the degassing of radioactive fluids. Depending on its origin, the gaseous radioactive waste is: • either filtered(1) and released into the atmosphere via the discharge stack. This is generally the case for gaseous waste coming from ventilation circuits; • or retained in the treatment system to reduce the level of radioactivity and then filtered and released into the atmosphere via the discharge stack. This is the case for gases released by the degassing of primary cooling water. In all circumstances, gaseous releases are controlled and monitored at the stack in order to check that these discharges do not have a noticeable impact on the terrestrial environment.
(1) Filtration allows retention of more than 99% of aerosols and iodines and their conversion into solid waste.
A concern having a bearing on the whole life cycle of the plant.
Protection of the public and of the environment requires optimization at each step of the EPR™ life cycle of: ➡design; ➡construction; ➡operation; ➡dismantling and decommissioning. Saint-Alban nuclear power plant, France (1,300 MWe x 2).
I 45 ■ENVIRONMENTAL IMPACT
Liquid radioactive waste is placed in two categories depending Solid radioactive waste on its origin: Reduction in the volume of solid radioactive waste to lessen the • waste from the primary system, which contains activation products unit’s impact on the environment was one of the objectives adopted (cobalt, manganese, tritium, carbon-14, etc.) minor quantities of at the design stage. dissolved fission gases (xenon, iodine, etc.), fission products Spent fuel is removed for either reprocessing or storage and residual (caesium, etc.), and also chemical substances such as boric acid waste created is packaged, to ensure that radioactive matter is and lithium hydroxide. The chemical substances can be recycled; confined, in glass representing once conditioned 5 m3 per year • waste from the systems connected to the primary system, which of high activity long-lived(3) waste and in concrete packages makes up the rest of the effluents. Among these, there are: representing once conditioned 4 m3 per year of long-lived – effluents which are radioactive and free from chemical pollution, intermediate level waste. – radioactive and chemically charged effluents, effluents with a very low level of radioactivity collected by the floor drains(2). The production of operating waste is limited to 80 m3 per year per unit thanks to the materials chosen to build the reactor and, After being systematically collected, this waste is treated to retain the zoning(4) of premises after start-up to reduce the possible most of its radioactivity. It is then channelled to storage tanks where it contamination of conventional equipment by radioactive material. undergoes both radioactive and chemical testing before being Short-lived low level waste, a by-product of the operating process, disposed of. is sorted, treated and stored in the EPR™ unit’s Radioactive Waste Processing Building so as to reduce the volume of waste as much as possible (compaction) and to ensure the radioactive material is confined through suitable packing.
(2) The floor drains constitute a network of underground pipes which collect material leaks, water from drainage operations, and water used to wash the floors. (3) Lifetime of a radioactive substance depends on its radioactive half-life period, which is the time after which its activity has reduced by a half because of the natural decrease of the radioactive source. (4) Separation of zones presenting contamination risks from other zones by means of doors, chambers, etc.
Typical distribution of annual dose to population from all sources
63% radon, soil and terrestrial materials 15% medical 12% background radiation (sun and Milky Way) 9.5% human body 0.49% miscellaneous artificial sources Sand sampling in the Cotentin peninsula (Flamanville and La Hague area). 0.01% nuclear power plants
46 I Chemical waste DECOMMISSIONING Operating a nuclear power station also involves the discharge of On the basis of the experience feedback resulting from dismantling liquid chemical waste and gaseous emissions. These wastes are operations performed in various countries on first generation nuclear processed, tested and monitored to ensure that their discharge is in power plants, the EPR™ unit’s design includes various measures compliance with authorized limits. which: • minimize the volume of radioactive structures, Conventional waste • reduce the potential hazard of the waste, for instance with the material choice minimizing hazardous substances, Conventional waste produced during the operating stage is also • lower the irradiation level of components submitted to fuel radiation, subjected to a strict management procedure so as to reduce its • restrict the spread of contamination and favor systems volume. It is sorted and stored in an adapted temporary location on decontamination, for example with the implementation of a the site. radiological zoning, Waste produced during the construction phase is treated in the • facilitate the access of personnel and machines and the evacuation same manner as during the operating phase, in a temporary location of waste, for instance with the implementation of suitable areas specifically created on the building site. The materials extracted and openings, during the earth-moving works, excavated material and rocks, are • ensure the gathering of building and operating data needed to reused as much as possible on the site as fill and in the making of prepare dismantling correctly. concrete after crushing. These measures facilitate the dismantling of the reactor to a level equivalent to IAEA Level 3 (return of site to common industrial use), limit the radiation doses of the corresponding operations and limit the quantity and activity of the nuclear waste produced.
IMPACT ON PUBLIC HEALTH
The impact of liquid and gaseous radioactive waste For the maximum gaseous and liquid radioactive waste discharges on public health cannot be measured directly. emissions from the entire site, calculating the impact Radionuclides introduced into the environment by the on health, for each inhabitant in the reference group, operation of a nuclear unit cannot be discerned by practical produces an annual effective dose amounting to a few tens means. Thus, the impact on public health must be assessed of microsieverts. theoretically by estimating the effective dose received In reality the nuclear units have a lower activity discharge by a hypothetical group of people, known as the reference than the maximum activity defined by this theoretical group. This group is considered as the group which would method. Considering the radioactive waste produced by be subject to the maximum effects of the gaseous and actual units, the impact on public health, for each member liquid waste if members remained permanently in residence (whether adult or infant) of the reference group, produces and only consumed local produce, and seafood fished an annual effective dose which is four to five times lower at the waste outlets. than the maximum values.
I 47 > EPR™ REACTOR Key to power station cutaway
1 Reactor building: 26 Safeguard building, inner and outer shell division 3 2 Polar crane 27 Emergency feedwater pump, division 3 3 Containment heat removal system: sprinklers 28 Medium head safety injection pump, division 3 4 Equipment hatch 29 5 Safeguard building, Refuelling machine division 4 6 Steam generator 30 Switchgear, division 4 7 Main steam lines 31 I & C cabinets 8 Main feedwater lines 32 Battery rooms, division 4 9 Control rod drives 33 Emergency feedwater 10 Reactor pressure vessel storage, division 4 11 Reactor coolant pump 34 CCWS heat exchanger, division 4 12 Reactor coolant piping 35 13 Low head safety injection CVCS heat exchanger pump, division 4 14 Corium spreading area 36 Component cooling water 15 In-containment refuelling surge tank, division 4 water storage tank 37 Containment heat removal 16 Residual heat removal system pump, division 4 system, heat exchanger 38 Containment heat removal 17 Safety injection system heat exchanger, accumulator tank division 4 39 18 Pressuriser Fuel building 40 19 Main steam isolation Fuel building crane valves 41 Spent fuel pool bridge 20 Feedwater valves 42 Spent fuel pool and fuel 21 Main steam safety and transfer pool relief valve exhaust 43 Fuel transfer tube silencer 44 Spent fuel pool cooler 22 Safeguard building division 2 45 Spent fuel pool cooling pump 23 Main Control Room 46 Nuclear auxiliary 24 Computer room building 25 Emergency feedwater 47 CVCS pump storage, division 2 48 Boric acid tank 49 Delay bed 50 Coolant storage tank 51 Vent stack
48 I
All over the world, AREVA provides its customers with solutions for carbon-free power generation and electricity transmission. With its knowledge and expertise in these fields, the group has a leading role to play in meeting the world’s energy needs. All rights All reserved. It Ranked first in the global nuclear power industry, AREVA's unique integrated offering covers every stage of the fuel cycle, reactor design and construction, and related services. statements The Group. AREVA In addition, the group is developing a portfolio of operations performance or warranties y in renewable energies. AREVA is also a world leader in electricity transmission and distribution and offers its customers a complete range of solutions for greater grid stability and energy efficiency.
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